ML19093A950

Submittal of Reponses to NRC Request for Clarification of Response to Request for Additional Information for the Technical Review of the Three Mile Island Unit 2 Independent Spent Fuel Storage Installation License Renewal Application

ML19093A950
Person / Time
Site:	Three Mile Island, 07200020
Issue date:	04/01/2019
From:	Wahnschaffe S US Dept of Energy, Idaho Operations Office
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards
References
CLN190779
Download: ML19093A950 (21)

Text

Department of Energy Idaho Operations Office 1955 Fremont Avenue Idaho Falls, ID 83415 April 1, 2019 Attn: Document Control Desk Director, Division of Spent. Fuel Management Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Submittal of Reponses to NRC Request for Clarification of Response to Request for Additional Information for the Technical Review of the Three Mile Island Uriit 2 Independent Spent Fuel Storage Installation License Renewal Application, Docket 72-0020 (CLN190779) .

REFERENCES:

1) Letter: DOE-ID EM~NRC-17-007 to NRC,

Subject:

Submittal of Application for Renewal of TMI-2 ISFSI License..SNM-2508, dated March 6, 2017

2) NRC Letter: Mr. John P. Zimmerman, Deputy Manager, Idaho Cleanup Project, Request for Clarification of Response to Request for Additional Information for Technical Review of the Application for Renewal of the Three Mile Island Unit 2 Independent Spent Fuel Storage Installation License No. SNM-2508 (CAC/EPID NOS. 001028/L-2017-RNW-0019 AND.000993/L-2017-RNE-0019), dated February 5, 2019

Dear Sir or Madam:

On March 6, 2017, the Department of Energy, Idaho Operations Office (DOE-ID) sµbmitted a license renewal application (LRA) requesting a 20-year renewal of Three Mile Island Unit 2 (TMI-2) Independent Spent Fuel Storage Installation (ISFSI) specific license SNM-2508 (Reference 1) (ADAMS Accession Nos. ML17089A501 andML17075Al99 through MLl 7075A201 ). The Nuclear Regulatory Commission (NRC) acknowledged acceptance of the LRA on May 5, 2017 (ADAMS Accession No. MLl 7125A284).

On February 5, 2019 DOE-ID received from the NRC a request for clarification of our response to the Request Additional Information (ADAMS Accession No. ML18360A186) pertaining to the technical review associated with the TMI-2 ISFSI LRA (Reference 2).

• Enclosure 1 to this letter contains the NRC requests for clarification and provides the DOE-ID responses. Also enclosed (enclosure 2) with this submittal is Revision 2 of the LRA, which includes changes that reflect the responses to the requests for clarification. Additionally, enclosures 3 and 4 contain the revised calculations referenced in enclosure 1. /V Iv/ .5S z. D A)i-l sszl,o

Document Control Desk CLN190779 Should you have questions or require additional information, please contact me at (208) 526-4993, or Steve Ahrendts at (208) 526-8888.

jf~

Steven Wahnschaffe License Manager

Enclosures:

1) Responses to NRC Request for Clarification of Response to Request for Additional Information on the TMI-2 ISFSI License Renewal Application

2) TMI-2 ISFSI License Renewal Application, Revision 2

3) Orano Federal Services Calculation CALC-3021323, "TMI-2 Canister Top-End Dose Rates forTMI-2 ISFSI License Renewal," Revision 1

4) Orano Federal Services Calculation CALC-3021788, "TMI-2 Canister Licon Criticality Analysis for TMI-2 ISFSI License Renewal," Revision 1 cc: Kristina Banovac, NRC Bernard White, NRC Meraj Rahimi, NRC Nicholas DiNunzio, DOE HQ Greg Sosson, DOE HQ

CLNI90779 Page 1 of 19 Responses to NRC Request for Clarification of Response to Request for Additional Information on the TMI-2 ISFSI License Renewal Application RAI 2-2 and RAI 3-6 Follow-up Revise the supplemental shielding analyses in support of excluding the dry shielded canister (DSC) basket from the scope ofrenewal review (Orano Federal Services Calculation CALC-3021323) to exclude the presence of the purge port shield block.

The supplemental shielding analyses in support of the exclusion of the DSC basket from the scope of renewal review rely on the presence of the purge port shield block. These analyses only look at dose rates for the vent port and filter hom;ing because, as stated in the reports documenting the analyses, the purge port dose rates are bounded by the vent port dose rates (e.g., see Section 3 .2, item 2 of LRA reference 3 .11.216). That can only be so because of the presence of the purge port shield block, which like the shield block for the vent port, ensures against radiation streaming through the purge port and that a TMI-2 canister cannot be located directly beneath the purge port.

However, the purge port shield block is scoped out of the renewal. Thus, the supplemental shielding analyses should include analyses for the purge port without credit for its shield block (in terms of both the shielding contribution an~ the relative placement of the TMI-2 canisters to the purge port). Alternatively, include the shield block in the renewal scope, provide an aging management review and any necessary corresponding time-limited aging analyses or aging management programs, and revise the UFSAR Supplement in Appendix C of the LRA, as appropriate.

This information is needed to determine compliance with Title 10 of the Code of Federal Regulations (10 CFR) 72.24(e), 72.122(h)(5), and 72.42(a).

• RAI 2-2 and RAI 3-6 Follow-up Response In response to the NRC staffs follow-up question regarding RAI 2~2 and 3-6, the supplemental shielding analysis, Orano Federal Services Calculation CALC-3 021323 (LRA Reference 3 .11.216), has been revised, adding an Appendix A. CALC-3021323 has been revised, modelling the purge port opening in the DSC while specifically not including the purge port block (Item #4 on Drawing 219-02-1003). Therefore, this revision col)siders the license renewal as-scoped configuration for both the vent .and purge port block components (i.e., present for the vent port shield block (scoped in) and absent for the purge port block (scoped out)). Reiterating the RA12-2 response, CALC-3021323 was performed using as a basis the approved FSAR design-basis shielding model, Calculation 219-02.0403, "Top End Dose Rates for TMI-2 Canisters" (LRA Reference 2.4.6). Therefore, CALC-3021323 maintains the existing ISFSI design basis configuration. It follows that both CALC-3021323 and Calculation 219-02.0403 are used solely in the design basis for assessing dose rates on the HEPA filter housing (i.e., UFSAR Table 7.3-1). For these dose rate assessments, the relevant dose rate acceptance criterion on the HEPA filter housing either for the vent or purge port is the Limiting Condition for Operation (LCO) 3.2.2 limit of 1200 mrem/hour. This acceptance limit is used for verifying compliance in CALC-3021323:

CLN190779 Page 2 of 19 Appendix A of the revised Calculation CALC-3021323 adds three new MCNP models. The rationale for including the three models is to answer unequivocally the follow-up clarification question. As a result, the response is broken down into three distinct aspects answered:

1. It is shown that the purge port block is not in scope.

2. The Justified Assumption 3.2.2 in CALC-3021323 is validated regarding the vent port (with its block) maximum dose rates bounding the purge port (without its block) dose rates.

3. There is not an increase, but rather a reduction in dose rates when the TMI-2 Canister clustering is varied to be in direct alignment with the purge port opening in the DSC.

The three MCNP .shielding models are all configured in the most conservative fashion, taking into account the RAI 2-2/3-6 combined scenario of: no Licon present, no DSC basket present, and the 12 TMI-2 Canisters closely clustered around the vent/purge port openings in the DSC.

The three MCNP shielding models are identified as follows:

Model 1: Modified vent port model (modified in comparison with the Licon and Basket RemovedzConfiguration

1. 2" model.in Revision O of CALC-3021323) with internal vent port shield block (Item #6 on Drawing 219-02-1002) present in Configuration #2 having the vent port shield block modeled as steel. The TMI-2 Fuel Canisters are clustered in Configuration #2 around the vent port shield block (See Figure A-1 in CALC-3021323).

Model 2: Purge port model with internal purge.port block (Item #4.on Drawing 219-02-1003) removed, in Configuration #2 and having the purge port block modeled as void. Credit was not taken for the shielding contribution from the purge port block, but credit was taken for its placement of the TMI-2 Fuel Canisters (i.e., space claim). Thus, the TMI-2 Fuel Canisters are arranged in Configuration #2 around the internal purge port block space (See Figure A-2 in CALC-3021323).

Model 3: Purge port model with internal purge port block removed in Configuration #3 having the purge port plock modeled as void and the TMI-2 Fuel Canisters arranged into Configuration #3. In Model 3, the TMI-2 Fuel Canisters occupy the space where the purge port block normally would be, whereas in Model 2 they did not. In this way, Model 3 does not take credit for the relative placement of the TMI-2 Fuel Canisters with respect to the purge port opening in the DSC (See Figure A-3 in CALC-3021323).

For information, "Configuration #2" and "Configuration #3" in the above model descriptions refer to two TMI-2 Canister reconfiguration scenarios. In both scenarios, the 12 TMI-2 Canisters are radially packed as closely as possible around the vent/purge port openings in the DSC and shifted axially toward the top-end in order to maximize dose rates. Configuration #2 is the worst-case TMI-2 Fuel Canister reconfiguration when credit is taken for positioning due to the purge/vent block space claim, while Configuration #3 is a reconfiguration removing the space claim of the purge/vent blocks. Configuration #2 is identified in Figure 5-4 of CALC-3021323, while Configuration #3 is identified in Figure A-3 of the same calculation.

CLN190779 Page 3 of 19 In order to answer the first aspect regarding scoping of the purge port block, Model 3 has the purge port block removed from the shielding model and the TMI-2 Fuel Canisters reconfigured around the purge port opening in t~e DSC. The model calculates a dose rate at the center of the purge port HEPA filter housing. The dose rate value remains below the LCO 3.2.2 limit of 1200 mrem/hour (i.e., maximum dose rate of256 mrein/hr). This supports the conclusion that the purge port block does not perform a credited shielding function, and is correctly scoped out of license renewal and no aging management review is necessary. In addition, as a point of clarity, the description of"purge port shield block" was inadvertently used in one location in CALC-3021323, and is now corrected in Revision 1 to be consistently identified as the "purge port block".

In order to answer the second aspect regarding validation of the original CALC-3021323 Justified Assumption 3 .2.2, a comparison was made of maximum dose rates between the Model 3 purge port without its internal purge port block present to the Model 1 vent port with its internal vent port shielding block present. Both Models 1 and 3 for the vent and purge ports, respectively, were evaluated with the 12 TMl-2 Canisters clustered around the vent/purge port openings in the DSC (i.e., Configuration #2 for the vent port shielding block with TMI-2 Canisters clustered around the vent port block (Model 1), and Configuration #3 for the purge port without the purge port ,

block, having the TMI-2 Canisters clustered around the purge port opening in the DSC (Model 3)). Through these shielding models, the original CALC-3021323 Justified Assumption 3.2.2 is verified. This is because the vent port maximum dose rates with its shielding block present (314 mrem/hr at edge and 265 mrem/hr in center of HEPA filter) bound the maximum purge port dose rates without its. purge port block present (256 mrem/hr).

Finally, the third aspect is answered regarding the relative placement of the 12 clustered TMI-2 Canisters with respect to the purge port opening in the DSC. A third shielding.configuration was run (Model 2) having the TMI-2 Canisters rotated slightly outward, as if the internal purge port block.was still present (A hypothetical as-modeled condition, but otherwise not physically possible). When comparing Model 2 to Model 3,it was confirmed that the TMI-2 Canister relative radial placement with respect to the purge port opening in the DSC is not relied upon in maintaining compliance with the 1200 mrem/hr dose rate limit. Model 3 has.the TMI-2 Canisters more closely packed around the purge port opening in the DSC (i.e, direct axial alignment), but actually shows a sizeable drop in the dose rates when compared to Model 2. Model 3 has lower dose rates due to this more direct axial alignment of the TMI-2 Canisters with the purge port opening in the DSC. This occurs because when the TMI-2 Canisters are in closer axial alignment to the purge port opening in the DSC, the TMI-2 Canisters' upper heads block direct radiation shine through the opening. The upper heads shield the more significant radiation coming through the sidewalls of the TMI-2 Canister shell and prevent streaming through the purge port opening in the DSC. As a result, this reduces the calculated dose rate in Model 3 to 25 6 mrem/hr at the center of the purge port HEPA filter housing when compared to the hypothetical Model 2 dose rate of 1121 mrem/hr.

. CLN190779 Attachment 1 Page 4 of 19 In conclusion, the 1200 mrem/hr LCO 3.2.2 limiting dose rate is maintained at the purge port HEPA filter housing, c~lculated under the extreme configuration of: the interior purge port block missing, the DSC basket removed, each individual TMI-2 Fuel Canister's Licon removed, and the 12 TMI:*2 Canisters clustered around the purge port opening in the DSC. Therefore, it can be concluded that the internal purge port block (Item #4 on Drawing 219-02-1003) is correctly scoped out of license renewal and no aging management review is necessary. In addition, by modeling shielding configurations without the purge port block and having the clustered TMI-2 Canisters both directly and indirectly aligned with the purge port opening m. the DSC, it is confirmed that compliance with the LCO 3,2.2 acceptance limit" does not depend on the relative placement between the TMI-2 Canisters and the purge port opening in the DSC. Furthermore, it also is concluded that corresponding dose rates at the vent port with the vent port shielding block present bound comparable ones at the purge port with the purge port block absent, validating CALC-3021323, Justified Assumption 3.2.2. LRA Section 3.8.4.5 and LRA Reference 3.11.216 have been updated as a result of these RAl follow-up clarifications.

RAI 2-2, 3-4 and 3-6 Follow-Up Provide the following additional information to clarify or support the responses to RAis 2-2, 3-4, and 3-6.

a) An eva\uation of dose rates at the horizontal storage module (HSM) rear a_ccess panel surface as a result of neglecting the TMI-2 fuel canister licon and the DSC basket's positioning of the TMI-2.

canisters within the DSC (relative to the DSC vent and purge ports).

The combined effects of neglecting the licon in the TMI-2 fuel canisters and neglecting the positioning of the canisters in the DSC by the DSC's basket resulted in calculated dose rates at the DSC vent filter housing of about 175 mrem/hr. Given this result and the trend in dose rates in the vicinity of the housing to be larger than the dose rates on the housing surface (indica~ed in LRA references 3 .11.102 and 3 .11.103 and by DOE-ID in the December 6, 2018, meeting), it is not clear that the dose rate limit for the HSM rear access door that is specified in the technical specifications would not be exceeded. Thus, the evaluation of

. the dose rate impacts should include dose rates at the HSM rear access panel surface.

b) Provide an evaluation of occupational doses for operations for DSCs without a basket and*

TMI-2 fuel canisters without licon that demonstrates occupational doses will remain within, or consistent with, the design basis described in UFSAR Section 7.4.1 and other relevant UFSAR sections (e.g., Section 7.1.2) and occupational dose limits in 10 CFR Part 20.

CLN190779 Page 5 of 19 The design bases in the UFSAR include an evaluation of occupational doses (see Section 7.4.1 of the UFSAR), which is based upon calculated dose rates from the UFSAR (see Section 7.3.2.2) and, for top end dose rates, appears to be based on LRA reference 2.4.6. Those calculated dose rates are in turn based on the DSC basket maintaining the positions of the TMI-2 canisters relative to each other and the DSC vent and purge ports. The dose rates are also based on the presence of the licon in the TMI-2 fuel canisters. To support the applicant's proposal to scope the DSC basket and TMI-2 fuel canister's licon out of renewal, the applicant conducted a supplemental analysis to evaluate the effects of neglecting the DSC's basket and TMI-2 fuel canister's licon on shielding and radiation protection. The applicant's supplemental analysis resulted in higher calculated dose rates for the top end areas of the DSC (i.e., 175 mrem/hr at the DSC vent filter housing calculated in the supplemental analysis, compared to

- 3 mrem/hr calculated in LRA reference 2.4.6). Thus, it is not clear that the design basis occupational dose analysis in the UFSAR would be met or maintained for DSCs in this configuration (i.e., no DSC basket and no TMI-2 fuel canister licon).

The applicant's supplemental analysis did include some evaluation with regard to compliance with 10 CFR Part 20 occupational dose limits for DSCs in this configuration; however, the evaluation is limited in the activities it considered and is limited to a single DSC. The evaluation should consider relevant transfer operations, periodic maintenance activities involving or around the vent and purge port filter housings and the HSM rear access door, and activities required by technical specifications (e.g., surveillances for limiting conditions for operation 3.1.1 and 3.2.3). The evaluation should also account for operations within a given year involving not just one, but multiple DSCs, up to all 29 loaded DSCs. Applying the calculated doses in the current supplemental analysis to all 29 DSCs indicates regulatory limits would be exceeded were a single individual to perform these activities for all 29 DSCs.

Thus, the evaluation should describe the actions that would be taken, controls that would be imposed, or conditions that would assure that occupational dose limits will not be exceeded. Guidance provided in Section 11.4.3.1 ofNUREG-1567, particularly the bulleted list at the end of the section, should be considered.

c) An explanation of the differences between the estimated dose rates for the vent port housing and the measured surface dose rates reported in LRA references 3 .11.102 and 3 .11.103.

It is not clear why measured dose rates on the vent filter housing surface, as reported in LRA references 3.11.102 and 3.11.103 (i.e., up to 15 mrem/hr), are higher than the dose rates estimated for that location in LRA reference 2.4.6 (i.e., - 3 mrem/hr).

d) An explanation of the locations of the measured 1 to 5 mrem/hr neutron dose rates reported in the text of the "Results" sections in LRA references 3.11.102 and 3.11.103.

The text of the "Results" section of LRA references 3 .11.102 and 3 .11.103 also states neutron dose rate measurements of 1 to 5 mrem/hr for HSMs 4 and 22; however, the location of these measured neutron dose rates (e.g., on the HSM rear access doors or the DSC purge and vent port filter housings) is unclear.

CLN190779 Page 6 of 19 This information is needed to determine compliance with 10 CFR 72.24(e), 72.122(h)(5), and 72.42(a).

RAI 2-2, 3-4 and 3-6 Follow-Up Response a) An evaluation of the dose rate at the Horizontal Storage Module (HSM) rear access panel surface as a result of neglecting the TMI-2 Fuel Canister Licon and the DSC basket's positioning of the TMI-2 Canisters within the DSC is provided in the original design basis calculation 219-02.0401, "HSM and Cask Dose Rate Calculation" (LRA Reference 3 .11.157). Table 11 in Calculation 219-02.0401 is reproduced as TMI-2 ISFSI FSAR Table 7.3-1, tabulating peak dose rate results on the HSM "Rear Wall," among other locations. In the table, the peak gamma dose rate of 104.5 mrem/hr corresponds to a position on the axial centerline of the DSC at the HSM outer rear wall surface. This dose rate location is 102 inches above the ISFSI basemat and corresponds to the large peak on the graph shown in Figure 5 of Calculation 219-02.0401. Of note, on page 16 of Calculation 219-02.0401 it is stated, "the vent dose rates will be over-estimated" because the source is smeared into a "homogenous cylindrical region" over the entirety of the DSC cavity.

From View N-N of drawing 219-02-6000, "Horizontal Storage Module Safety Analysis Report,"

the top edge of the door is 8.56 inches below this 102-in. elevation (i.e., 93.44 inches above the ISFSI basemat). The bottom of the door extends down another 26 inches from this location (i.e., 67.44 inches above the ISFSI basemat). As mentioned above, Figure 5 in Calculation 219-02.0401 shows a graph of the dose rate with respect to the height above the ISFSI basemat (solid line on graph represents the rear HSM wall). Of note, as stated on page 17 of Calculation 219-02.0401, "The 1.5 inch steel cover is assumed to be recessed into the module to simplify the model", indicating the rear access door was modelled :flush with the HSM rear wall. From the Figure 5 in Calculation 219-02.0401, in the range of 67.44 to 93 .44 inches above the ISFSI basemat, the maximum dose rate is approximately 32 mrem/hour on the rear access door. The measured dose rates at. all of the rear access doors in 2010 and 2012 (per LRA Reference 3 .11.102 and 3.11.103), were all less than 1 mrem/hour. The Technical Specification dose rate limit for the HSM rear access door is 100 mrem/hour (LCO 3.2.2). Therefore, the LCO dose rate limit for the HSM rear access door will not be exceeded as demonstrated by the original design basis calculation and actual measurement data.

CLN190779 Page 7 of 19 b) An evaluation of the occupational dose rates as a result of neglecting the TMI-2 Fuel*Canister Licon and the DSC basket's positioning of the TMI-2 Canisters within the DSC has previously been provided in the original design basis Calculation 219-02.0402, "INEL/TMI-2 ISFSI Site Dose Rate Calculation." Calculation 219-02.0402 uses the design basis DSC/HSM models in Calculation 219-02.0401, which neglect both the Licon and DSC basket (see response to sub-question (a) above) in calculating dose rates on the surface of the HSM. Based on this input, Calculation 219-02.0402 estimates occupational dose rates in the vicinity surrounding the TMI-2 ISFSI. The dose rates from this calculation are reported in the TMI-2 ISFSI UFSAR design bases, including in Section 7.4, Table 7.4-2, .and Figure 7.4-2. Because the existing design bases for the TMI-2 ISFSI are unchanged, the occupational dose analysis already contained in the UFSAR is met for DSCs in the configuration without a DSC basket and without Licon in the TMI-2 Fuel Canister.

The supplemental analysis included in the RAI responses (CALC-3021323) was completed in response to the various RAI questions. This analysis does not alter or supersede the existing design bases as described in th~ UFSAR (i.e., no new design configuration or altered activities from those described in thilnfsAR were generated). As such, the design basis described in UFSAR Section 7.4 already-co,vers 10 CFR 20 occupational dose limits for the operational configurations of the TMI-2.IS_FSI, with identified loading activities previously tabulated in Table 7.4-1.

As described in UFSAR Ch~p\er 7, a Radiation Protection Program (RPP) is implemented to control work activities in radia!ion areas, including periodic maintenance activities. The implementing procedure is-S-'f~-NLF-RAD-001, "TMI-2 Radiation Protection Program." This procedure ensures occupationaJ exposures at the TMI-2 ISFSI (other than planned special exposures) are well below reg1Jlatory limits by adhering to an administrative individual dose limit of twenty percent of the limits;specified in 10 CFR 20.1201. Occupational exposures are tracked in accordance with MCP-188, tTLD Usage and Obtaining Personnel Dose History."

Calculated dose rates used in lrcensing design bases are, by definition, conservatively high because they use bounding source terms and assumptions. The conservatively-calculated dose rates are used for licensing so that an occupational exposure estimate may be made to provide reasonable assurance that the 10 ~FR 20 occupational exposure limits will not be exceeded during ISFSI operations. During actual ISFSI operations, calculated dose rates are not used to demonstrate compliance with 10 CFR 20 limits. Work planning that includes radiation surveys of actual dose rates, use of temporary shielding (if appropriate), control of stay times, radiation work permits (RWPs), and personnel dose rate and accumulated dose monitoring is used to ensure regulatory dose limits are not exceeded by any individual.

CLN190779 Page 8 of 19 Requirements for RWPs at the TMI-2 ISFSI are included in the operating procedures. If during work planning, it is determined that an RWP is required, it is included in the procedure or work order directing the work in the field. The R WP documents current radiological conditions for the work location and evaluates points for which work is paused and further radiation condition evaluation is required before work is allowed to proceed. The R WP process 'also includes limiting radiation conditions (e.g., dose rate and/or fixed or airborne contamination levels) that void the RWP and cause a "Stop Work" condition.

At the TMI-2 ISFSI, any work activity involving opening the HSM rear access door requires an RWP and full-time coverage from a radiological control technician (RCT). The area inside the HSM rear access door is surveyed for the actual dose rate and posted as a radiological area. Steps are included in the procedure for the RCT to perform monitoring to determine radiological conditions prior to allowing any work inside the area to proceed. All work is required by the RPP to be performed in accordance with the R WP. The current radiation dose rates inside and around the HSM rear access door are so low that personnel electronic dosimetry is not required when performing these tasks. This situation is expected to continue through the ISFSI PEO due to the continuous decay of the radioactive source term inside the TMI-2 Canisters. However, if Radiation Control personnel were to find elevated dose rates, they would have the option to require electronic dosimetry that would alarm.

c) One reason the measured dose rates in LRA references 3 .11.102 and 3 .11.103 appear to be higher than those in LRA Reference 2.4.6 is due to MCNP modelling detector error percentages within the original design basis calculation (LRA Reference 2.4.6). The MCNP error percentage determined for the -3mrem/hr value in the original MCNP model is 19% (as indicated in Table 6-1 of CALC-3021323). According to the MCNP manual, detector result errors between 10% and 20% are "questionable", whereas below 5% are "generally reliable". When the dose rate was recomputed with a 5% error in CALC-3021323, the value increased to 35 mrem/hr, bounding the measured values in LRA references 3 .11.102 and 3 .11.103.

Another reason for a lower dose rate value in the original design bases calculation over the LRA references 3 .11.102 and 3 .11.103 was due to geometric modelling differences from actual hardware. Originally, the MCNP results in LRA Reference 2.4.6 were computed for a transport/sampling configuration (with the sample cover). Modelling this cover provided additional shielding that does not exist in the storage configuration. Such modelling parameters were maintained during CALC-3021323 development so that no change to the design basis was introduced.

d) The locations of the measured 1 to 5 mrem/hr neutron dose rates (for HSMs 4 and 22) reported in the text of the "Results" sections in LRA references 3 .11.102 and 3 .11.103 was answered in response to RAI 3-4, sub-questions (a) and (b). The response to sub-question (a) indicated that the dose rate was on the HEPA filter housing surface while the response to sub-question (b) specifically placed the location near the geometric center of the filter housing.

CLN190779 Page 9 of 19 RAI 2-5 Follow-Up Propose a license condition or technical specification to use transfer cask spacers that are aged less than 20 years, and revise the UFSAR Supplement in Appendix C of the LRA, as appropriate.

The current TMI-2 design bases include the use of transfer cask spacers, as discussed in the UFSAR.

Also, while the UFSAR shielding analyses do not credit the material of the transfer cask spacers, the analyses do credit the spacers' function of axially positioning the DSC within the transfer cask. The positioning of the DSC within the transfer cask has an important effect on transfer cask dose rates, both axially and radially for areas where the transfer casks' radial shielding changes. To address this effect and to support scoping out the transfer cask spacers, the applicant provided a supplemental analysis to evaluate the dose rate effects of neglecting the spacers to demonstrate that regulatory requirements would still be met when accounting for those effects. In addition, the applicant also noted in the RAI response that the spacers used during the initial ISFSI loading campaign no longer exist, and new spacers would need to be fabricated if the transfer cask were used in the future at the ISFSI (e.g., for retrieval operations from storage).

At the December 6, 2018, meeting, the NRC staff indicated it had questions regarding the supplemental .

analysis, and DOE-ID and NRC staff discussed the approach of a license condition limiting the age of spacers, given new spacers would need to be fabricated for future use. Under this approach, the spacers would scope into the renewal; however, the license condition would preclude the need to conduct an aging management review of the spacers.

The applicant should also include any corresponding changes to UFSAR Supplement in Appendix C of the LRA (e.g., scoping tables). The NRC staff is amenable to discussing alternative ~pproaches upon DOE-ID request.

This information is needed to determine compliance with 10 CFR 72.42(a)

RAI 2-5 Follow-Up Response To address the concerns regarding RAI 2-5, DOE-ID is proposing License Condition 21 to account for the TC spacers. License Condition 21 will ensure TC spacers aged less than 20 years are used at the TMI-2 ISFSI. DOE-ID recognizes that the original TC spacers used during the initial ISFSI loading campaign no longer exist; therefore, new TC spacers would need to be fabricated at the time they are required.

Because the original TC spacers no longer exist and a license condition is now proposed, there will be no TC spacers exceeding 20 years of age that will perform an intended function during the PEO. It therefore follows that classifying the spacers as out-of-scope for renewal is consistent with Section 2.4.3 of NUREG-1927 for replaceable components, in which "other active components/systems" are "subject to a change in configuration or replacement based on a qualified life or service." Therefore, no changes to the LRA scoping tables are required because the TC spacers are bundled with the other out-of-scope Handling and Transfer Equipment addressed in LRA Section 2.3.3.2. The text in LRA Section 2.3.3.2 has been revised to incorporate these changes to the TC spacers and Appe,ndix D.2.2 includes the new license condition. Conforming FSAR changes in Sections C.2.6.3 and C.3 are incorporated to reflect this new license condition.

CLN190779 Attachment 1 Page 10 of 19 RAI 3-5 Follow-Up Revise the proposed UFSAR Supplement in Appendix C of the LRA to ensure that the HSM AMP adequately addresses the potential for aggregate reactions (alkali silica reaction, ASR).

In RAJ 3-5, the staff requested that the applicant justify its conclusion that aging effects due to ASR in the HSM concrete are not credible. In its response, the applicant provided additional details of the petrographic characterization of cores obtained from HSMs affected by early design-related degradation.

These results showed evidence of ASR (isolated small patches of white ASR gel) observed in one of six core samples taken during this evaluation in 2009. However, the applicant still concluded ASR to not *be credible due to the limited identification of reactive aggregates per these results, the concrete mix specification used for HSM fabrication, and the limited presence of water during operations.

During the December 6, 2018, meeting, the NRC staff noted that it does not agree with the justification for excluding aggregate reactions as a credible aging mechanism, per the technical basis provided in Section 3_.5.1.3 ofNUREG-2214 (Managing Aging Processes In Storage (MAPS) Report), currently in final publication review. In NUREG-2214, the staff states that ASR, the most common aggregate reaction, is generally a slow degradation mechanism. ASR may take from 3 to more than 25 years to develop in concrete structures, depending on the nature (reactivity level) of the aggregates, the moisture and temperature conditions to which the structures are exposed, and the concrete alkali content. The delay in exhibiting deterioration indicates that there may be less reactive forms of silica that can eventually cause deterioration.

Operating experience has revealed degradation of the concrete in the Seabrook reactor containment as a result of ASR (see NRC Information Notice (IN) 2011-20, ADAMS Accession No. MLl 12241029). The concrete used at the Seabrook nuclear power plant passed all industry standard ASR screening tests at the time of construction. However, ASR-induced degradation was identified in August 2010. Per IN 2011-20, licensees that tested using American Society for Testing and Materials (ASTM) C227 and ASTM C289 could have concrete that is susceptible to ASR-induced degradation since these standard methods may not accurately predict aggregate reactivity when dealing with late- or slow-expanding aggregates containing strained quartz or microcrystalline quartz. In addition, ASR screening tests are generally not conducted on each aggregate source but rather in select batches, which increases the risk for use of aggregates of different reactivities when procured from different sources.

Due to the uncertainties in screening tests that can effectively be used to eliminate the potential for ASR and previous ASR operating experience at a nuclear facility, the staff considers the aging mechanism to be credible in concrete exposed to any environment with available moisture, and therefore, aging management is required during the 40-year timeframe. In its response to RAJ 3-5, the applicant did not provide a technical basis to support that the conclusions in IN 2011-20 are not applicable to the TMI-2 HSM concrete (i.e., that the petrographic test methods used to assess aggregate reactivity of the TMI-2 HSM concrete aggregates is not subject to the uncertainties and risks of underestimated reactivities as discussed in IN 2011-20).

CLN190779 Attachment 1 Page 11 of 19 In the December 6, 2018, meeting, DOE-ID and the NRC staff disc~ssed options to address the staffs concern and how the final approach should be reflected in the design bases (i.e., the UFSAR Supplement to be .incorporated upon license renewal). In one approach, the applicant may choose to expand the scope of the HSM AMP to include aging effects due to aggregate reactions (i.e., define these aging effects to be credible). In an alternative approach, the applicant may credit the use of the American Concrete Institute (ACI) 349.3R second tier acceptance criteria in the HSM AMP but predefine a corrective action that ensures that conditions not meeting these acceptance criteria will be evaluated to ensure ASR is not the

• apparent or root cause. Per NUREG-2214, the NRC staff considers the use of these acceptance criteria to be acceptable for managing ASR aging effects. However, the NRC staff wants to ensure that the HSM AMP for the TMI-2 ISFSI properly captures the potential for ASR. The NRC staff is amenable to discussing further alternative approaches upon DOE-ID request.

This information is needed to determine compliance with 10 CFR 72.42(a).

RAI 3-5 Follow-Up Response DOE"'.'ID has reviewed the staffs RAI 3-5 follow-up question regarding Alkali Silica Reaction (ASR) ...

Although DOE-ID still concludes that it is unlikely that ASR is a relevant aging mechanism for the TMI-2 ISFSI site, DOE-ID is including. a required corrective action activity in the HSM AMP to investigate if ASR is the apparent or root cause initiator of any degradation exceeding ACI 349 .JR, second-tier acceptance criteria. This "'."as a suggested alternative approach to handling the concern in the follow-up question. Changes are reflected in updates to LRA Section 3.5.4.2.5, LRA Appendix A2.7, and LRA Table A-2. For further context, DOE-ID is providing additional information below regarding the RAI 3-5

. follow-up question and the cited operating experience (OE) as justification for the conclusion that the likelihood of an ASR aging mechanism for concrete at the TMI-2 ISFSI site is very low.

DOE-ID has reviewed NRC Information Notice (IN) 2011-20 pertaining to ASR occurrences at the

• Seabrook power station. DOE-ID finds that this OE is not completely relevant to the TMI-2 ISFSI site because the environmental conditions necessary for ASR formation and progression between the Seabrook site and the TMI-2 ISFSI site diverge considerably. NextEra (the Seabrook licensee) performed an extent of condition (EOC) review regar~ing ASR (ADAMS Accession No. ML16224B079),

concluding, "Because the ASR mechanism requires the presence of moisture or very high humidity in the concrete, ASR has been predominantly detected in portions of pelow-grade structures, with limited impact to exterior surfaces of above grade structures. "

CLN190779 Attachment 1 Page 12 of 19 ASR-affected Seabrook concrete Was initially found in 2009 within below-grade tunnel structures, and was subsequently found primarily in other below-grade locations (ADAMS Accession No.

ML12151A397). Seabrook concrete*structures are laid out in an arrangement with large vertical dimensions. Many of the Seabrook structures have underground areas at least 40 feet below grade, while other structures have underground areas that are 80 feet below grade. As evidenced by core samples, groundwater infiltration at Seabrook in the below grade walls of concrete structures had caused loss of compressive strength and modulus of elasticity in the concrete from ASR. The NextEra report stated the groundwater conditions as follows, "The results showed some of the groundwater to be aggressive.

Ground water testing performed in November 2008 and September 2009 found pH values -between 6. 0 I and 7.51, chloride'values between 19 ppm and 3900 ppm, and sulfate values between JO ppm and JOO ppm. Aggressive chemical attack becomes a concern when environmental conditions exceed threshold values (Chlorides> 500 ppm, Sulfates> 1500 ppm, or pH< 5.5)." None of the TMI-2 ISFSI HSMs are located below grade, immersed in a soil environment or have any of these required chemical conditions.

Therefore, aggressive forms of chemical attack are not applicable for the concrete HSMs at the TMI-2 ISFSI (see LRA Section 3.5.4.2.2).

In addition, Seabrook is located directly on a salt-water body (the Atlantic Ocean), while the TMI-2 ISFSI is located in a high, dry desert plateau. Notwithstanding this primary ocean effect, and the potential for salt-water chlorides initiating ASR as a prevailing factor at Seabrook for above-grade concrete, the NextEra EOC still concluded that ASR was, predominantly detected in portions of below-grade structures".

DOE-ID agrees that both the TMI-2 ISFSI site and the Seabrook site share similarities in that the concrete used at both locations was not originally expected to be susceptible to ASR during construction due to the fo,llowing: (1) the coarse aggregate passed the ASR reactivity testing and (2) low-alkali cement.

(<0.60% total alkali) was used. While it is not definitively conclusive that the TMI-2 ISFSI aggregate.

used in the* HSM concrete is completely non-reactive, aggregate ASR reactivity is only one of three required contributing factors necessary for ASR propagation. As mentioned in both the RAI 3-5 response and in revised LRA Section 3.5.4.2.5, one of three required elements for ASR progression is adequate moisture. This is consistent with Section 3.5.1.JofNUREG -2214, which states that the minimum threshold is, "available moisture, generally accepted to be relative humidity greater than 80 percent

. (Pedneault, 1996; Stark, 1991)." As stated in LRA Section 3.5.3, the maximum ambient relative humidity at the TMI-2 ISFSI is 70%, which is below the 80% relative threshold that is needed to provide the "requisite conditions for initiation and propagation ofASR" (NUREG-2214) .. As such, this characteristic is not present at the TMI-2 ISFSI site to support this aging mechanism.

CLN190779 Page 13 of 19 Nonetheless, DOE-ID recognizes the NRC staffs concerns regarding possible unknown causal factors involving ASR. Consequently, DOE-ID has revised the corrective action program associated with the HSM AMP. DOE-ID has expanded the discussion of the corrective action program element in Appendix A2. 7 of the LRA, using the acceptance criteria defined in element Appendix A2.6. l as any degradation exceeding second-tier criteria, as requiring a technical evaluation. DOE-ID has included a specific required corrective action for any HSM concrete subject to inspection in AMP Table A-2. This corrective action adds an initiating contributor check, looking for ASR as either an apparent or root cause of the .

premature degradation.

We note that in preparing the response to this question that LRA Tables A-1 and A-2 for the DSC and HSM AMPs, respectively, would benefit from additional clarification. Specifically, additional clarification has been provided pertaining to due dates for the required baseline and AMP inspections and with respect to the timing of the license renewal effective date, including the addition of an appropriate grace period. The tables have been revised accordingly.

RAI 3-7 and RAI 3-8 Follow-up Modify the benchmark analysis for the criticality calculations that support the responses to RAis 3-7 and 3-8 to include sufficient trending with respect to key parameters and to include experiments with materials relevant to the criticality calculations. Also, confirm that the benchmark models used the same modeling techniques as the TMI-2 fuel canister analysis models.

Benchmark analyses for typical criticality evaluations consider trends on several parameters. However, the benchmark analysis for the calculations supporting the RAI responses only looks at trends for two parameters: the energy of average lethargy of fission and the uranium-235 number density in the fuel.

While the calculations are not for a typical case, trends with respect to other parameters should have been considered, such as the ratio of hydrogen to fissile atoms. Also, the selected benchmark experiments include experiments with materials that are not relevant to the TMI-2 fuel canisters, such as cadmium.

The benchmark analysis should include, to the extent practical, only experiments with materials that are in the analyzed system (i.e., the TMI-2 fuel canisters in the DSC in an HSM). Thus, the evaluation of trends in the benchmark analysis should determine the trends when experiments with the non-relevant materials are excluded to ensure an appropriate or bounding bias and bias uncertainty is used in the criticality evaluation.

In addition, the models for the benchmark analysis should use the same techniques as were used in the fuel canister analysis. This includes use of the cellmix option to create the material used in the geometry to represent the fuel both with and without water from the cell data card information as opposed to explicit modeling of the fuel pellets in the geometry. If the model for the benchmark analysis did not use the same techniques used in the fuel canister analysis, the applicant should revise the benchmark analysis accordingly.

This information is needed to determine compliance with 10 CPR 72.42,(a).

CLN190779 Page 14 of 19 RAI 3-7 and RAI 3-8 Follow-up Response The benchmark analysis contained within the RAI 3-7 and 3-8 response supplementary criticality evaluation (Orano Federal Services Calculation CALC-3021788, LRA Reference 3.11.226) has been revised to incorporate additional trending parameters, a more restrictive benchmark set, and usage of the same modeling techniques in all benchmarks as were used in the TMI-2 Fuel Canister models. The Upper .

Subcritical Limit (USL) has been*recalculated and it is confirmed that the TMI-2 Fuel Canisters will not exceed the updated USL.

The original benchmark analysis-in CALC-3021788, Revision 0, examined only two trending parameters, Energy of Average Lethargy of Fission (EALF) and U-235 number density in the fuel, while the revised benchmark analysis adds two new trending parameters, atom ratio of hydrogen to U-235 and enrichment.

Additionally, to further ensure that the selected experiments adequately represent the TMI-2 Fuel Canister criticality cases, a "reduced" benchmark set is analyzed alongside reanalysis of the original benchmark set The reduced benchmark set removes all benchmark cases in the original benchmark set that contain reflector or absorber materials not relevant to the TMI-2 Fuel Canister criticality cases. Such materials include lead, Boraflex 1, copper, and cadmium. The reduced benchmark set contains 38 benchmark cases from the original benchmark set of 50 benchmark cases. All benchmarks are revis~d to incorporate usage of the cellmix option to be consistent with the TMI-2 Fuel Canister criticality cases (no other inconsistent techniques were utilized in the original analysis).

Eight USL values are calculated using both the original and reduced benchmark sets with each of the four trending parameters. To be conservative, the minimum USL value calculated is applied as the sub-criticality a~ceptance criteria. The minimum USL value in the revised calculation is 0.9435. The maximum calculated TMI-2 Fuel Canister keffvalue is 0.85926. Thus, it is confirmed that the TMI-2 Fuel Canister will maintain criticality safety following revision of the benchmark analysis.

1 Described as "Boroflex" in associated benchmark literature (Experiment LEU-COMP-THERM-042, International

.

Handbook ofEvaluated .

Criticality Safety Benchmark Experiments, Nuclear Energy Agency, NEAINSC/DOC(95)03, September 2015)

CLN190779 Page 15 of 19 RAI 3-9 Follow-Up Provide a clear and concise justification that the maximum water content assumed in the criticality analyses in the FSAR and in the supplementary criticality analysis (in response to RAis 3-7 and 3-8) is not, or will not, be exceeded.

The maximum amount of water assumed in the analyses includes bound and unbound water that remained in the TMI-2 canisters after drying and water that could be reacquired during storage at the ISFSI, ,

including for the requested period of extended operation. It is still not clear from the response to RAI 3-9 and the supporting references that the drying of the TMI-2 canisters ensured this maximum amount was not exceeded. In particular, LRA reference 3.11.5 discusses different conditions of drying (e.g., specific temperatures at specific pressures), which if met would ensure a certain level of dryness, or maximum residual bound and unbound water. It also appears to discuss different processes or indications (e.g.,

looking at plateaus in falling drying rates). However, the reference also appears to indicate that, in different instances, different criteria and indications were used for different canisters in ways that do not seem to be consistent. Also, while some canisters may have been checked for dryness, it is not clear how that assured all canisters were at the same level of dryness or did not exceed the residual water maximum amount. The licensee should provide a clear description of how the canisters were actually verified to not exceed the appropriate residual (bound and unbound) water maximum amount (including for the licon for the fuel canister since water here can get into the fuel cavity), the criteria that were used, and clear confirmation that the canisters were verified to meet those criteria. Any applicable references or appropriate supporting documentation should also be provided.

This information is needed to determine compliance with 10 CFR 72.42(a).

RAI 3-9 Follow-Up Response As described in response to RAI 3-9, 8.0 liters is the maximum assumed design basis amount of water within the fuel region of the TMI-2 Canister allowed by the TMI-2 ISFSI second FSAR criticality analysis, INEEL/INT-99-00126 (LRA Reference 3.11.106) and the RAI 3-7/3-8 supplementary criticality analysis, CALC-3021788 (LRA Reference 3.11.226). This maximum design basis quantity of water bounds the maximum hypothetically calculated allowance limit of 3.2 liters water in each TMI-2 Canister.

As stated in LRA Section 3.3.1.2.1, this 3.2-liter quantity in each TMI-2 Canister includes three sources:

bound, unbound, and reacquired water. This 3.2-liter value consists of2.3 liters of combined bound and unbound water, as referenced from EDF-1466 (LRA Reference 3.11.5) and 0.9 liters ofreacquired water, as referenced from EDF-797 (LRA Reference 3.11.104). As detailed below, the 2.3-liter volume of water is not exceeded and this fact is validated by process controls for the heated-vacuum drying (HVD) process as recorded in the implementing procedures for drying.

CLN190779 Attachment 1 Page 16 of 19 Licon is only present in the TMI-2 Fuel Canisters. In response to the RAI 3-9 question on preferential Licon water reabsorption (sub-question 'b'), an increase of the 0.9-liter quantity ofreacquired water to a maximum of2.58 liters in each of the four TMI-2 Fuel Canisters in DSC-16 was conservatively calculated. Under this conservative scenario developed in response to the RAI, the maximum water in the fuel region of each TMI-2 Fuel Canister in DSC-16 could be as much as 4.88 liters (2.3 liters of bound and unbound water plus 2.58 liters ofreacquired water). As stated above, up to 8.0 liters of water can be present in the fuel region of each TMI-2 Canister while maintaining criticality safety (TMI-2 ISFSI FSAR, Section 3.3.4.4.1). The supplementary criticality analysis, CALC-3021788 also confirmed this 8.0-liter water limit was an acceptable volumetric quantity for maintaining sub-criticality, even when the Licon was removed from the model. Since both the design basis maximum water content of 3 .2 liters and the RAI 3-9 calculated maximum water content of 4.88 liters are bounded by the criticality limit of 8.0 liters, the maximum water content assumed in the TMI-2 ISFSI FSAR criticality analyses and in CALC-3021788 will not be exceeded.

In order to ensure water within the TMI-2 Canisters met these design limits,.specifically the 2.3 liter combined bound/unbound water limit, each TMI-2 Canister was vacuumed-dried and heated simultaneously (i.e., heated-vacuum dried (HVD)). It should be noted that the unbound water limit makes up 0.17-liter of the total 2.3-liter water limit. The origin of this design basis unbound water limit of 0.17 liters was the FSAR original criticality evaluation (FSAR Section 3.3.4.1), as previously described in the RAI 3-9(a) response.

Section 5.1.1.1 of the TMI-2 ISFSI FSAR provides the basis for the TMI-2 Canister drying process acceptance criterion, stating the following: "Prior to placement in dry storage, the TMI-2 canisters are to be dried to ensure that no free water is contained in the canisters. A verified record of final TMI-2 canister drying will be maintained for each canister." Given the HVD process was performed in accordance with procedures developed under quality assurance (QA) controls, the actual bound and unbound water content in any of the TMI-2 Canisters complies with the design basis limit of2.3 liters.

This maximum limit was not exceeded in any TMI-2 Canister during the HVD process by ensuring that each documented drying cycle met three drying acceptance criteria (EDF-'1466, Section 7.4), "1) the system is operating properly; 2) the unbound water is removed; and 3) core debris temperature has been raised above 170°F", indicating the bound water has been sufficiently removed.

At the Test Area North (TAN) Hot Shop, there were two Heated Vacuum Drying Systems (HVDS) that dried up to four TMI-2 Canisters at a time (as documented in TPR-1190 for HVDS #1 and TPR-6596 for HVDS #2) [TPR-1190, "TMI-2 Canister Drying" (LRA Reference 3.11.97) and TPR-6596, "HVDS-2 TMI-2 Canister Drying"]. The drying procedures were written to. ensure compliance with the three drying acceptance criteria and then it followed that each TMI-2 Canister drying cycle procedurally demonstrated that these three EDF-1466 acceptance criteria were met. As such, each TMI-2 Canister has a documented record of the drying cycle being performed, indicating each TMI-2 Canister was verified to comply with the three drying acceptance criteria in EDF-1466. TPR-1190 (LRA Reference 3.11.97) provides a sample procedural record of the drying campaign for four TMI-2 Fuel Canisters (Serial Numbers D-281, D-280, D-211, and D-173).

CLN190779 Page 17 of 19 Appendix I in TPR-1190, Revision 15 has the steps demonstrating the procedural basis for the TMI-2 ISFSI FSAR, Section 5.1.1.1 dryness requirements and provides the implementing criteria (i.e., Steps 2.8, 3.4.10, 4.3.10.8, 4.3.11, 4.3.13.1, 4.3.14.1, 4.7.7.8.1, 4.5.18, and Appendix E, Dryness Acceptance Definition). Given each drying cycle procedural QA record, these particular steps (or similar depending on procedure revision) and the sign-offs for each provides the documented QA record that all three EDF-1466, Section 7.4 dryness acceptance criteria were met. Because each set of four TMI-2 Canisters subject to the HVD process would have included varied contents (some would contain partially intact fuel assemblies, some TMI-2 Filter Canister fines, others portions ofloose rubble), each drying cycle would have been necessarily unique. For example, near intact or partially intact fuel debris may represent drying cycles with smaller water volumes, with unbound water removed more quickly than drying cycles with higher water content. Therefore, measuring the change in slope of the initial falling rate drying may be more conducive using the second method in EDF-1466, Section 7.4.2.2 (see below). However, all of the drying cycles maintained compliant procedural steps demonstrating dryness acceptability. DOE-ID Records Management maintains the QA records for all the TMI-2 Canisters' drying procedures, providing dryness acceptance evidence from each batch ofTMI-2 Canisters dried. An explanation below describes how the TMI-2 Canisters were verified to stay below the appropriate residual (bound and unbound) water maximum amount and the basis used for that determination.

For the first EDF-1466 dryness acceptance criterion, operating controls ensured the extent ofTMI-2 core debris drying would meet the 2.3-liter design water limit. These controls were then implemented for each TMI-2 Canister via the above-mentioned TAN drying procedures. EDF-1466 clearly indicates in Section 7.4.1 that the system operating parameters provided in TPR-1190, Revision 15 ensure adequate dryness (e.g., these parameters included system instrument calibration, that thermocouples remained operational during the drying cycle, and that isolation valves were operable).

For the second EDF-1466 dryness acceptance criterion regarding unbound water, plotting the "falling rate" drying curve for each drying cycle was used to determine when the unbound water remaining in the TMI-2 core debris was removed. Falling rate drying refers to the boiling rate of the water declining as the "critical water content" has been reached (i.e., water has been boiled away). During falling rate drying, the pressurization rate in the furnace is proportional to the amount of water remaining in the HVDS.

The HVDS heaters evaporated the liquid water and generated the pressure by which dryness (using the "falling rate" criteria) was verified. Dryness was validated by periodically isolating the furnace and measuring the rate of furnace pressure increase. The relative change in the rate of pressure increase during drying isolations was used to evaluate the remaining water content by determining when the drying had entered into "falling rate" criteria. For the unbound water, drying proceeds until the falling rate curve meets applicable acceptance criteria using one of the three methods outlined in Section 7.4.2.2 of EDF-1466. These methods demonstrate acceptance using a specified decline in the falling rate value or by the rate declining below a maximum. The TAN drying procedures specify and implement the EDF-1466 (Section 7.4.2.2) falling rate drying methodology for dryness acceptance (e.g., Appendix I, "Procedure Basis" in TPR-1190, Revision 15).

CLN190779 Page 18 of 19 For the third EDF-1466 dryness acceptance criterion regarding bound water, the maximum TMI-2 core debris temperature is the most significant parameter for controlling the amount of bound water remaining.

As stated above, EDF-1466 requires that TMI-2 core debris temperatures must increase above 170°F in order to reduce the bound water to an acceptable volume. The TMI-2 core debris was heated sufficiently long after the bulk of the unbound water was removed, ensuring that the TMI-2 core debris temperature exceeded 170°F. As shown in Figure 6 ofEDF-1466, this 170°F temperature corresponds to a 2.13-liter unbound water design volume (2.3 L-0.17 L = 2.13 L).

Because TMI-2 core debris temperatures cannot be measured directly, the results of a heat transfer model for the drying cycle runs are used to infer that the minimum 170°F temperature has been exceeded, by examining both the temperature values at the furnace thermocouple locations and the time to reach those temperatures. That heat transfer model was evaluated in EDF-1469, "Temperature Benchmarking for Dryer Runs TMI-005 and TMI-009," as described in the response to RAJ 3-9. This thermal model of the HVDS evaluated the TMI-2. core debris temperature by providing temperature conditions throughout the HVDS furnace, including nodes representative of the various thermocouple locations. EDF-1469 demonstrated conservative correspondence between model and actual run conditions. Heater operation was verified when furnace temperatures had increased above specified values in TPR-1190 for HVDS #1 and TPR-6596 for HVDS #2. The heating time provides additional assurance of dryness and is confirmed using EDF-1469 benchmarking data. EDF-1469 calculated the maximum TMI-2 core debris temperatures for system operation out to 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, representing actual drying cycle HVDC-009 which was run for over 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br />.

CLN190779 Page 19 of 19 In accordance with Section 7.4.3 ofEDF-1466, "the debris temperature may be estimated if the following two parameters are known; 1) the heating rate of the debds after the unbound water is removed; and

2) the debris temperature during the removal of the unbound water." In order to estimate these two parameters, the model ofHVDC-009 developed in EDF-1469 provides TMI-2 core debris temperatures.

This data from EDF-1469 is shown in Figure 23 ofEDF-1466. The figure shows the temperature rise is minimal for the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> as the rapidly evaporating water removes the heat. The temperature subsequently increases uniformly with the heat addition rate. The break in the slope at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in Figure 23 indicates the onset of initial falling rate drying. This 8-hour period envelopes the removal of unbound water, allowing the TMI-2 core debris temperature to increase and freeing up bound water.

Figure 22 in EDF-1466 indicates that falling rate drying proceeds until approximately 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br />. As shown in Figure 23 ofEDF-1466, at approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> (near midway point of falling rate drying -

between 8 and 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br />), the TMI-2 core debris temperature is 350°F. As such, the temperature rise for an HVDS furnace load similar to that in run HVDC-009 at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> operation will be more than double the minimum 170°F temperature required to control the bound water to acceptable volumetric limits*

(i.e., 2.13 liters bound water). At approximately a minimum of9 hours of heating, Figure 23 shows a TMI-2 core debris temperature of 170°F, thus meeting this volumetric bound water limit. As stated above, the drying procedures (TPR-1190 for HVDS #1 and TPR-6596 for HVDS #2) have discrete steps requiring signatures, in order to satisfy these dryness acceptance requirements. For example, in establishing the thermocouple measurements were adequate to support the above temperature/heating criteria, Step 4.3.11.A in TPR-1190, Revision 15, stated that any of the four thermocouples maintained a minimum 650°F temperature for at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

In summary, the above description provides clear implementing procedural controls and QA records for HVD of the TMI-2 Canisters and resultant dryness assurance. These implementing procedural controls validate that the 8.0-liter maximum allowed water content assumed in the second FSAR criticality analysis and in CALC-3021788 bounds the design basis water content allowance for each TMI-2 Canister of 3.2 liters total, of which 2.3 liters comes from the bound/unbound water limits. As a separate confirmation of the negligible water content in the system, measurement data has indicated miniscule and steady-state DSC hydrogen concentrations over the years (See LRA Figure 3-6), indicating the small hydrogen concentrations are not being generated via radiolysis of water.